The following relates to the nuclear reactor arts, electrical power generation arts, nuclear reactor control arts, nuclear electrical power generation control arts, and related arts.
A pressurized water reactor (PWR) employs a pressure vessel containing subcooled water as the primary coolant. Hot, subcooled water is circulated between the reactor core and one or more steam generators to transfer energy from the reactor core to the steam generator. In a conventional design, the steam generators are separate elements and the primary coolant is coupled between the pressure vessel and the steam generator via suitable high pressure fluid conduits. In an integral PWR design, the one or more steam generators are located inside the pressure vessel.
An electrically heated pressurizer is used to control the reactor coolant system pressure. The PWR contains a steam region that controls the pressure and changes its volume to accommodate changes in liquid volume in the pressure vessel and coolant loop(s). A combination of heater operation and subcooled water spray in the steam region is used to increase or decrease the amount of steam in a steam “bubble” to maintain the pressure vessel at a constant pressure. It is desired to suppress or prevent over-pressurization of the PWR during operating transients and to preserve the subcooled margin of the reactor coolant inside the reactor core to ensure adequate heat transfer.
In a typical PWR control paradigm, the water level in the pressurizer (that is, the pressurizer water level) serves as an indication of reactor coolant inventory. If the pressurizer water level drops below a predetermined level, additional makeup water is pumped into the reactor coolant system (RCS). On the other hand, if the pressurizer water level rises above another preset limit, water is letdown from the RCS.
Reactor power control in a PWR is typically a complex process in which numerous control variables (for example, steam flow, feedwater flow, “gray” control rod positioning) are concurrently adjusted to maintain the desired operational state and to control transient behavior. These adjustments are constrained by the requirement to keep the pressurizer level within the preset limits so as to avoid overpressurization or underpressurization.
To provide additional PWR reactivity control, it is known to add an effective amount of a soluble neutron poison, typically boron in the form of boric acid, to the primary coolant. The soluble boron poison reduces the magnitude of the moderator temperature coefficient. Thus, the boron concentration in the primary coolant provides yet another “adjustment knob” for controlling reactor power output. In some PWR's, the concentration of boric acid is varied over the fuel cycle to offset changes in reactor core reactivity as the nuclear fuel is consumed. The boron concentration is selected such that the moderator temperature coefficient and moderator void coefficient both remain negative. The magnitude of these coefficients is substantially reduced by the addition of the boron poison; however as a result, changes in water temperature inside the reactor have reduced impact on core power, thus simplifying control.
With reference to FIG. 1, a typical relationship between the reactor coolant temperature (abscissa) and reactor power (ordinate) is shown. FIG. 1 plots the hot leg temperature (Thot) relationship, the cold leg temperature (Tcold) relationship, and the average temperature (Tave) relationship. The integrated control system simultaneously controls steam flow, feedwater flow, and the control rods in the reactor to alter core power output for transients. The soluble boron poison is typically reduced as the fuel burns to minimize the rod insertion into the reactor core. FIG. 1 plots the programmed reactor coolant temperatures inside the reactor pressure vessel using this approach. As power level increases, reactor inlet temperature drops and reactor outlet temperature rises to maintain a constant average temperature in the core and steam generator. The constant average temperature minimizes any volume changes in the reactor coolant as the power level changes. As a result, the pressurizer water level between the heated water and steam regions in the pressurizer remains essentially constant during changes in power level.
Any deviation of the pressurizer water level that does exceed the upper or lower preset level limit causes water to be letdown or makeup water added to the RCS, in order to maintain the constant pressurizer water level setpoint. However, such corrective events are not common during normal operation, because the moderator temperature coefficient and moderator void coefficient are both small, and the control system maintains a nearly constant average water temperature.
Disclosed herein are further improvements that provide reduced cost, simplified manufacturing, and other benefits that will become apparent to the skilled artisan upon reading the following.